Experimental evaluation of the water ingression mechanism for corium cooling

Experiments were performed to assess the significance of water ingression cooling in the quenching of molten corium. Water ingression is a mechanism by which water penetrates into cracks and pores of solidified corium to enhance cooling that would otherwise be severely limited by the low thermal conductivity of the material. Quench tests were conducted with 2100 °C melts weighing 75 kg composed of UO₂, ZrO₂ and chemical constituents of concrete. The amount of concrete in the melts was varied between 4% and 23%. The melts were quenched with an overlying water layer; three tests were conducted at a system pressure of 1 bar and four tests at 4 bar. The measured cooling rates were found to decrease with increasing concrete content and, contrary to expectations, are essentially independent of system pressure. For the lower concrete content melts, cooling rates exceeded the conduction-limited rate with the difference being attributed to the water ingression mechanism. Measurements of the permeability of the corium “ingots” produced by the quench tests were used to obtain a second, independent set of dryout heat flux data, which exhibits the same trend as the quench test data. The data was used to validate an existing dryout heat flux model based on corium permeability associated with thermally induced cracking. The model uses the thermal and mechanical properties of the corium and coolant, and it reproduces the very particular data trend found for the dryout heat flux as a function of concrete content. The model predicts that water ingression cooling would be most effective for concrete-free corium mixtures such as in-vessel type melts. For such a melt the model predicts a dryout heat flux of 400 kW/m² at a pressure of 1 bar. The results of this study provide an experimental basis for a water ingression model that can be incorporated into computer codes used to assess accident management strategies.     

Corium crust strength measurements

Corium strength is of interest in the context of a severe reactor accident in which molten core material melts through the reactor vessel and collects on the containment basemat. Some accident management strategies involve pouring water over the melt to solidify it and halt corium/concrete interactions. The effectiveness of this method could be influenced by the strength of the corium crust at the interface between the melt and coolant. A strong, coherent crust anchored to the containment walls could allow the yet-molten corium to fall away from the crust as it erodes the basemat, thereby thermally decoupling the melt from the coolant and sharply reducing the cooling rate. This paper presents a diverse collection of measurements of the mechanical strength of corium. The data is based on load tests of corium samples in three different contexts: (1) small blocks cut from the debris of the large-scale MACE experiments, (2) 30 cm-diameter, 75 kg ingots produced by SSWICS quench tests, and (3) high temperature crusts loaded during large-scale corium/concrete interaction (CCI) tests. In every case the corium consisted of varying proportions of UO₂, ZrO₂, and the constituents of concrete to represent a LWR melt at different stages of a molten core/concrete interaction. The collection of data was used to assess the strength and stability of an anchored, plant-scale crust. The results indicate that such a crust is likely to be too weak to support itself above the melt. It is therefore improbable that an anchored crust configuration could persist and the melt become thermally decoupled from the water layer to restrict cooling and prolong an attack of the reactor cavity concrete.     

A numerical and experimental study of water ingression phenomena in melt pool coolability

During a postulated severe accident, the core can melt and the melt can fail the reactor vessel. Subsequently, the molten corium can be relocated in the containment cavity forming a melt pool. The melt pool can be flooded with water at the top for quenching it. However, the question that arises is to what extent the water can ingress in the corium melt pool to cool and quench it. To reveal that, a numerical study has been carried out using the computer code MELCOOL. The code considers the heat transfer behaviour in axial and radial directions from the molten pool to the overlaying water, crust generation and growth, thermal stresses built-in the crust, disintegration of crust into debris, natural convection heat transfer in debris and water ingression into the debris bed. To validate the computer code, experiments were conducted in a facility named as core melt coolability (COMECO). The facility consists of a test section (200 mm × 200 mm square cross-section) and with a height of 300 mm. About 14 L of melt comprising of 30% CaO + 70% B₂O₃ (by wt.) was poured into the test section. The melt was heated by four heaters from outside the test section to simulate the decay heat of corium. The melt was water flooded from the top, and the depth of water pool was kept constant at around 700 mm throughout the experiment. The transient temperature behaviour in the melt pool at different axial and radial locations was measured with 24 K-type thermocouples and the steam flow rate was measured using a vortex flow meter. The melt temperature measurements indicated that water could ingress only up to a certain depth into the melt pool. The MELCOOL predictions were compared with the test data for the temperature distribution inside the molten pool. The code was found to simulate the quenching behaviour and depth of water ingression quite well.     

Ex-vessel core melt stabilization research (ECOSTAR)

The project on ex-vessel core melt stabilization research (ECOSTAR) started in January 2000 to be concluded by end of 2003. The project is performed by 14 partner institutions from five European countries and involves a large number of experiments with low- and high-temperature simulant melts and real corium at different scales. Model development and scaling analysis allows application of the research results to existing and to future LWRs in the area of reactor design and accident mitigation. The project is oriented toward the analysis and mitigation of severe accident sequences that could occur in the ex-vessel phase of a postulated core melt accident. The issues are: (1) the release of melt form the pressure vessel, (2) the transfer and spreading of the melt on the basement, (3) the analysis of the physical–chemical processes that are important for corium behavior especially during concrete erosion with onset of solidification, and (4) stabilization of the melt by cooling through direct water contact. The results achieved so far resolve a number of important issues: the amount of melt that could be transferred at RPV failure from the RPV into the containment can be substantially reduced by lowering the residual pressure in the primary circuit. It is found that melt dispersion also strongly depends on the location of the RPV failure, and that lateral failure results in substantially less melt dispersion. During melt release, the impinging melt jet could erode parts of the upper basement surface. Jet experiments and a derived heat transfer relation allow estimation of its contribution to concrete erosion. Spreading of the corium melt on the available basement surface is an important process, which defines the initial conditions for concrete attack or for the efficiency of cooling in case of water contact, respectively. Validation of the spreading codes based on a large-scale benchmark experiment is underway and will allow determination of the initial conditions, for which a corium melt can be assumed to spread homogeneously over the available surface. Experiments with UO₂-based corium melts highlight the role of phase segregation during onset of melt solidification and during concrete erosion. To cool the spread corium melt, the efficacy of top flooding and bottom flooding is investigated in small-scale and in large-scale experiments, supported by model developments. Project assessment is continuing to apply the results to present and future reactors.     

Large-scale ECOKATS experiments: Spreading of oxide melt on ceramic and concrete surfaces

Large-scale ECOKATS experiments are performed to study spreading of an oxide melt on ceramic and concrete surfaces. The oxide melt generated by a thermite reaction was composed of 41 wt.% Al₂O₃, 24 wt.% FeO, 19 wt.% CaO and 16 wt.% SiO₂. This melt was selected as the most appropriate simulation of a corium melt because of its wide freezing range of approx. 450 K. Despite a rather low liquidus temperature, the attempt to measure melt viscosity failed. As spreading of high-temperature oxide melts is nearly isothermal during the early phase of motion, i.e., only thin thermal boundary layers will develop, the melt viscosity can be estimated from a two-dimensional spreading experiment, ECOKATS-V1, on a ceramic substrate by approximate self-similar solutions. To further study the influence of the gas release from the substrate caused by thermal erosion of the underlying concrete by a corium melt on spreading, a large amount of the oxide melt was released into a 2.6 m long and 0.29 m wide channel leading into a 3 m × 4 m rectangular surface. Spreading on a concrete substrate is influenced by the gas release from the decomposed concrete, which changes viscosity. A viscosity increase by a factor of 3.6 was estimated from spreading in the concrete channel.     

Experiments on in-vessel melt coolability in the EC-FOREVER Program

This paper reports the results from the experiments conducted on the coolability of corium melt during a severe accident scenario when the bottom head is full of the core melt, undergoing natural circulation. These experiments are part of the EC-FOREVER Program in which vessel failure experiments have also been performed. The experiments are performed in a 1/10th scale vessel (400 mm diameter and 15 mm wall thickness) and the oxidic melt employed is the mixture CaO + B₂O₃ at 1400 K, representing the corium melt mixture of UO₂ + ZrO₂.The experiments employed an initial phase, during which uniform volumetric heating of the melt was provided and the vessel was pressurised to 25 bar, for several hours, to generate maximum creep deformation of 5%, in order to provide the conditions for the formation of a gap between the melt-pool crust and the bottom head wall. After this phase, the vessel was flooded with water.Data were obtained on only the vessel and the melt pool temperatures in one of the EC-FOREVER experiments reported here. In the second experiment, however, besides the temperature data, additional data were obtained on the steam flow rate and the heat transfer to the water, at the upper face of the melt pool, as a function of time.It was found that the gap cooling mechanism was not effective in reducing the vessel wall temperatures after water flooding. Post-test examinations revealed that the water ingression extended to the depth of only 60 mm in the melt pool. The character of the heat transfer to the water from the melt pool upper surface was found to be similar to that observed in the MACE tests for the coolability of an ex-vessel melt pool flooded by water at the top.     

Ex-vessel corium spreading: results from the VULCANO spreading tests

In the hypothetical case of a nuclear reactor severe accident, the reactor core could melt and form a mixture, called corium, of highly refractory oxides (UO₂, ZrO₂) and metallic or oxidized steel, that could eventually flow out of the vessel and mix with the basemat decomposition products (generally oxides such as SiO₂, Al₂O₃, CaO, Fe₂O₃, …). For some years, the French Atomic Energy Commission (CEA) has launched an R&D program which aimed at providing the tools for improving the mastering of severe accidents.Within this program, the VULCANO experimental facility is operated to perform experiments with prototypic corium (corium of realistic chemical composition including depleted UO₂). This is coupled with the use of specific high-temperature instrumentation requiring in situ cross calibration. This paper is devoted to the “spreading experiments” performed in the VULCANO facility, in which the effects of flow and solidification are studied.Due to the complex behavior of corium in the solidification range, an interdisciplinary approach has been used combining thermodynamics of multicomponent mixtures, rheological models of silicic semisolid materials, heat transfer at high temperatures, free-surface flow of a fluid with temperature-dependant properties.Twelve high-temperature spreading tests have been performed and analyzed. The main experimental results are the good spreadability of corium–concrete mixtures having large solidification ranges even with viscous silicic melts, the change of microstructure due to cooling rates, the occurrence of a large thermal contact resistance at the corium–substrate interface, the presence of a steep viscosity gradient at the surface, the transient concrete ablation. Furthermore, the experiments showed the presence of the gaseous inclusions in the melt even without concrete substrate. This gas release is linked to the local oxygen content in the melt which is function of the nature of the atmosphere, of the phases (FeO_x, UO_y, …) and of the substrate. These tests with prototypic material have improved our knowledge on corium and contributed to validate spreading models and codes which are used for the assessment of corium mastering concepts.     

Controlling and cooling core melts outside the pressure vessel

For future reactors, the control and cooling of ex-vessel corium melts is under consideration to increase the passive safety features even for very unlikely severe accidents. In this context, different research activities are studying ex-vessel corium behaviour and control, including the implementation of a core cooling device outside the reactor pressure vessel in order to prevent basement erosion and to maintain the integrity of the containment. This paper describes current research on key phenomena which must be understood and quantified to be finally controlled by the cooling device. These are the release of corium melt from the pressure vessel, the temporary retention of the melt in the reactor cavity until melt through of the gate, spreading of the melt on a large surface, and finally the cooling and solidification of the melt by direct water contact. The experiments use high temperature melts which are similar to corium melts. Where necessary, models are developed to transfer the results to reactor scale.     

Natural convection heat transfer test for in-vessel retention at prototypic Rayleigh numbers – Results of COPRA experiments

Large-scale COPRA experiments were performed to investigate the natural convection heat transfer in melt pools for the in-vessel retention during severe accidents in Chinese large-scale advanced PWRs. Both water and binary mixture of 20 mol% NaNO₃ – 80 mol% KNO₃ were used as the melt simulant material in performed tests. Due to the full scale geometry of the COPRA test section, the Rayleigh numbers of the melt pool could reach up to the prototypic magnitude of 10¹⁶. Natural convection heat transfer tests at prototypic Rayleigh numbers have been performed to study the influence of the heat generation rate and melt simulant material on the melt pool temperature, heat flux distribution and heat transfer capability. The comparisons of the melt pool temperature and heat flux distribution from water experiments and molten binary salt experiments showed that the crust formation along the inner surface of the vessel wall could impact the heat transfer characteristics of the melt pool. And the heat flux distribution from COPRA water tests and molten salt tests were in good agreement with those from Jahn-Reineke water experiments and RASPLAV molten salt experiments, respectively. The heat transfer capability of the melt pool Nu_dn from COPRA molten salt tests were larger than those from water tests, but both were lower than those from ACOPO and BALI predictions within the same range of Rayleigh numbers (10¹⁵ – 10¹⁷).     

In-vessel melt pool coolibility test—Description and results of LIVE experiments

The LIVE test program investigates in-vessel melt pool behaviour and cooling strategies for in-vessel corium retention during severe accidents in LWRs. The main part of the LIVE facility is a 1:5 scaled semi-spherical lower head of a typical pressurized water reactor. Up to now, LIVE experiments have been performed in different external cooling conditions, melt volumes and heat generation rates. At present the well-known simulant material KNO₃-NaNO₃ in non-eutectic composition (80 mole% KNO₃-20 mole% NaNO₃) and in eutectic composition (50 mole% KNO₃-50 mole% NaNO₃) is used. This work presents the behaviour of a homogenous melt pool regarding the 3D heat flux distribution through vessel wall, melt pool temperature, crust thickness and the pool melt composition in transient and in steady state conditions.     

Simulation of melt jet breakup and debris bed formation in water pools with IKEJET/IKEMIX

The objective of the development of the code system KESS is simulating the processes of core melting, relocation of core material to the lower head of the reactor pressure vessel (RPV) and its further heatup, modelling of fission product release and coolability of the core material. In the scope of the code development, IKEJET and IKEMIX were designed as key models for the breakup of a molten jet falling into a water pool, cooling of fragments and the formation of particulate debris beds. Calculations were performed with these codes, simulating FARO corium quenching experiments at saturated (L-28) and subcooled (L-31) conditions, as well as PREMIX experiments, e.g. PM-16. With the assumption of a reduced interfacial friction between water and steam as compared to usually applied laws, the melt breakup, energy release from the melt and pressurisation of the vessel observed in the experiments are reproduced with a reasonable accuracy. The model is further applied to reactor conditions, calculating the relocation of a mass of corium of 30 t into the lower plenum, its fragmentation and the formation of a particle bed.     

Solid particle effects on heat transfer in a multi-layered molten pool with gas injection

In the very unlikely event of a severe reactor accident involving core melt and pressure vessel failure, it is important to identify the circumstances that would allow the molten core material to cool down and resolidify, bringing core debris to a stable coolable state. To achieve this, it has been proposed to flood the cavity with water from above forming a layered structure where upward heat loss from the molten pool to the water will cause the core material to quench and solidify. In this situation the molten pool would become a three-phase mixture: e.g., a solid and liquid slurry formed by the molten pool as it cools to a temperature below the temperature of liquidus, agitated by the gases formed in the concrete ablation process. The present work quantifies the partition of the heat losses upward and downward in this multi-layered configuration, considering the influence of the viscosity and the solid fraction in the pool, from test data obtained from intermediate scale experiments at the University of Wisconsin-Madison. These experimental results show heat transfer behavior for multi-layered pools for a range of viscosities and solid fractions. These results are compared to previous experimental studies and well known correlations and models.     

Analytical tool development for coarse break-up of a molten jet in a deep water pool

A computer code JASMINE-pre was developed for the prediction of premixing conditions of fuel–coolant interactions and debris bed formation behavior relevant to severe accidents of light water reactors. In JASMINE-pre code, a melt model which consists of three components of sub-models for melt jet, melt particles and melt pool, is coupled with a two-phase flow model derived from ACE-3D code developed at JAERI. The melt jet and melt pool models are one-dimensional representations of a molten core stream falling into a water pool and a continuous melt body agglomerated on the bottom, respectively. The melt particles generated by the melt jet break-up are modeled based on a Lagrangian grouped particle concept. Additionally, a simplified model pmjet was developed which considers only steady state break-up of the melt jet, cooling and settlement of particles in a stationary water pool. The FARO corium quenching experiments with a saturation temperature water pool and a subcooled water pool were simulated with JASMINE-pre and pmjet. JASMINE-pre reproduced the pressurization and fragmentation behavior observed in the experiments with a reasonable accuracy. Also, the influences of model parameters on the pressurization and fragmentation were examined. The calculation results showed a quasi-steady state phase of melt jet break-up during which the amount of molten mass contained in the premixture was kept almost constant, and the steady state molten premixed masses evaluated by JASMINE-pre and pmjet agreed well.     

A use of prototypic material for the investigation of severe accident progression

An experimental research platform using corium melts is established for the understanding of safety related important phenomena during a severe accident progression. The research platform includes TROI facility for corium water interaction experiments and VESTA facility for corium-structural material interaction experiments. A cold crucible technology is adapted and improved for a generation of 5–100 kg of corium melts at various compositions. TROI facility is used for experiments to investigate premixing and explosion behaviors during a fuel coolant interaction process. More than 70 experiments using corium at various compositions were performed to simulate steam explosion phenomena in a reactor situation. The results indicate that the conversion efficiency of steam explosion for corium is less than 1%. VESTA facility is used to investigate molten corium-structural material interaction phenomena. VESTA facility consists of two cold crucibles. One crucible is used for the melting of charged material and pouring of corium melt. The other crucible is used for the corium-structural material interaction while providing an induction heating to simulate the decay heat. The results of an experiment on the interaction between corium melt and a specimen made of Inconel performed in the VESTA facility is reported.     

BETA experiments on melt-concrete interaction: the role of zirconium and the potential sump water contact during core melt-down accidents

In the BETA test facility of Kernforschungszentrum Karlsruhe, prototypical core melts can be simulated in concrete structures sufficient in size to allow a computer-code-assisted extrapolation to be made to the reactor geometry.Three experiments have been carried out to investigate special aspects of molten corium interacting with concrete. The investigations and measurements show the dominance of Zr oxidation during concrete attack by the chemical reduction of SiO₂ to elemental Si and the subsequent Si oxidation by the gases from the concrete.Additionally, the failure of a cylindrical concrete wall was studied, which is eroded on the inner side by a heated melt while being cooled outside by stagnant water. In the experiment wall failure occurs and the melt relocates into the water annulus.Application of the experimental results to light-water reactor severe accidents is discussed.     

Effect of melt water interaction configuration on the process of steam explosion

Steam explosion experiments are performed at various modes of melt water interaction configuration using prototypic corium melt. The tests are performed to simulate both melt water interaction in a partially flooded cavity and melt water interaction in a cavity with submerged reactor. The tests are performed using zirconia and corium melts. The behavior of melt jet fragmentation during the flight in the air and fragmentation and mixing of melt jet in water is investigated by a high-speed video visualization and by comparison of debris size distribution and morphology of debris. Strength of steam explosion is estimated by measuring dynamic pressure and dynamic force.     

Fuel coolant interaction experiments in TROI using a UO2/ZrO2 mixture

This paper discusses the results of steam explosion experiments using reactor material carried out under “Test for Real cOrium Interaction with water (TROI)” program. About 4–9 kg of corium melt jet is delivered into a sub-cooled water pool at atmospheric pressure. Spontaneous steam explosions are observed in four tests among six tests. The dynamic pressure, dynamic load, and morphology of debris clearly indicate the cases with steam explosion. The initial conditions and results of the experiments are discussed.     

Characteristics of corium debris bed generated in large-scale fuel-coolant interaction experiments

The formation of corium debris as the result of fuel-coolant interaction (energetic or not) has been studied experimentally in the FARO and KROTOS facilities operated at JRC-Ispra between 1991 and 1999. Experiments were performed with 3–177 kg of UO₂–ZrO₂ and UO₂–ZrO₂–Zr melts, quenched in water at depth between 1 and 2 m, and pressure between 0.1 and 5.0 MPa. The effect of various parameters such as melt composition, system pressure, water depth and subcooling on the quenching processes, debris characteristics and thermal load on bottom head were investigated, thus, giving a large palette of data for realistic reactor situations.Available data related to debris coolability aspects in particular are:

• Geometrical configuration of the collected debris.

• Partition between loose and agglomerated (“cake”) debris.

• Particle size distribution with and without energetic interaction.

These data are synthesised in the present contribution.     

Development and verification of molten corium–concrete interaction code

During a severe accident of Pressurized Water Reactor(PWR), the core materials was heated, melt located on the lower head of Reactor Pressure Vessel(RPV). With the temperature rise, the corium will melt through the lower head and discharge into the reactor cavity. Those corium will interact with the concrete and damage the integrity of the containment, so some coolability method should used to quench the corium. In order to investigate the progress of MCCI, a MCCI analysis code, that is MOCO, was developed. The MOCO includes the heat transfer behavior in axial and radial directions from the molten corium to the basemat and sidewall concrete, crust generation and growth, and coolability mechanisms reveal the concrete erosion and gas release, which are important for the interaction process. Cavity ablation depth, melt temperature, and gas release are the key parameters in the interaction research. The physical-chemistry reaction is also involved in MOCO code. In the present paper, the related MCCI experiment data were used to verify the models of the MOCO and the calculation results of MOCO were also compared with other MCCI analysis codes.